Zirconium metal alloys are widely used in core components and structures of water cooled nuclear fission reactors because of their low neutron cross section, among other apt properties for such service. Note for instance U.S. Pat. No. 4,212,686. Several zirconium alloy compositions have been developed and marketed primarily for nuclear reactor applications. Typical of such alloy compositions of zirconium are the commercially available materials identified as Zircaloy-2 and Zircaloy-4, comprising alloys set forth in U.S. Pat. Nos. 2,772,964 and 3,148,055. A niobium containing alloy of zirconium for reactor service is disclosed in U.S. Pat. Nos. 3,150,972 and 4,212,686.
The Zircaloys comprise alloy compositions containing at least about 95% by weight of zirconium metal and including in percent by weight up to about 2.0% of tin, up to about 0.5% of iron, up to about 0.5% of chromium and 0 to about 0.15% of nickel.
The degree of susceptibility to corrosion of a material is a critical factor regarding its use or performance in a water cooled reactor. In a reactor environment zirconium alloys normally form a relatively innocuous, dark surface oxide uniformly and superficially thereover. This so-called black oxide provides protection to the underlying metal and thickens with increased reactor residence at a slow rate. However, zirconium alloy can further develop deleterious nodules of corrosion, sometimes referred to as pustular corrosion. The nodular type of corrosion rapidly increases in size or area and depth over the alloy surface, which under certain conditions may impair the integrity of the alloy. Nodular corrosion comprises a white oxide that can grow several times faster than the innocuous black surface oxide to produce a thick white oxide layer impeding heat transfer among the other impediments.
The degree of susceptibility to nodular corrosion of zirconium alloys when exposed to the environment of a water cooled reactor has been found to be dependent upon several or a combustion of factors, including particular alloy composition and microstructure thereof, as well as the temperatures of the reactor in operation. See, for example, the disclosures of U.S. Pat. Nos. 3,150,972, 3,261,682 and 4,212,686.
As noted in U.S. Pat. No. 4,238,251, there is an evident correlation between microstructural characteristics of a zirconium alloy composition and resistance to nodular-type of corrosion in a reactor environment.
Manipulation of the microstructure of zirconium alloys through annealing procedures has been proposed in the art as a means for improving resistance to corrosion as well as enhancing other crucial properties of such alloys for use in reactor service. U.S. Pat. Nos. 2,736,651, 2,894,866 and 3,884,728, for instance, teach reforming of the microstructure of certain alloys of zirconium to increase their structural strength and corrosion resistance in reactor service.
However, the microstructure of zirconium alloys, and in turn their corrosion susceptibility, have been found to often vary. Different or nonuniform microstructures within an alloy can result from faulty or incomplete annealing, and from metal working or fabrication operations comprising reduction or drawing, shaping or cutting procedures, and welding.
Accordingly, there can be a great deal of latitude or uncertainty as to degree of susceptibility to corrosion for reactor components such as fuel cladding and channels which are formed from a zirconium alloy composition.
The disclosure of each of the foregoing cited United States Patents is incorporated herein by reference.